The Pressurized Water Reactor (PWR) 100, shown as FIG. 1, is the most common type of nuclear power generating reactor, with over 230 in use for power generation and a further several hundred in naval propulsion. It uses ordinary water as both coolant 118 and moderator 116. The design is distinguished by having a primary cooling circuit 102 which flows through the core 110 of the reactor 100 under very high pressure, and a secondary circuit 104 in which steam 114 is generated to drive the turbine generator 106. The core 110 and primary cooling circuit 102 are contained within a concrete containment structure 101.
A PWR 100 may have fuel assemblies 120 of 200-300 rods 108 each, arranged vertically in the core 110, and a large reactor may have about 150-250 fuel assemblies 120 with 80-100 tons of uranium. Water in the reactor core reaches about 325° C., hence it must be kept under about 150 times atmospheric pressure to prevent it from boiling. Pressure is maintained by steam in a pressure vessel 112. In the primary cooling circuit 102, which is circulated via primary-side pump 124, the water is also the moderator 116, and if any of it turned to steam the fission reaction would slow down. This negative feedback effect is one of the safety features of this type of reactor. The secondary shutdown system (not shown) involves adding boron to the primary circuit 102.
The secondary circuit 104 is under less pressure and the water there, being in thermal contact with the primary circuit 102, boils in the heat exchangers (not shown) within the steam generator 122. The steam drives the turbine generator 106 to produce electricity, and is then condensed and returned via a secondary-side pump 126 to the heat exchangers (not shown) within the steam generator 122.
Referring now to FIG. 2, the Boiling Water Reactor (BWR) 200 has many similarities to the PWR, except that there is only a single circuit 204, which passes through the concrete containment structure 202, in which the water is at lower pressure (about 75 times atmospheric pressure) so that it boils in the core 210 at about 285° C. The reactor 200 is designed to operate with 12-15% of the water in the top part of the core 210, which is housed within a pressure vessel 212, as steam 214, and hence with less moderating effect and efficiency than the PWR. The steam 214 passes through drier plates 228 (steam separators) above the core 210 and then directly to the turbines 206, which are thus part of the reactor circuit 204. The reactor circuit 204 also includes a core-circulating pump 224 to circulate the boiling water in the pressure vessel 212, and a recycle pump 226 which returns condensed steam 214 that has passed through the turbine 206 back to the pressure vessel 212.
A BWR fuel assembly 220 comprises 90-100 fuel rods 208, which are secured by control rods 230, and there are up to 750 assemblies 220 in a reactor core, holding up to 140 tons of uranium. The secondary control system (not shown) involves restricting water flow through the core so that steam in the top part means moderation is reduced.
During operation of a nuclear power reactor, impurities and products of the reactor coolant are deposited on nuclear fuel assemblies. These deposits can impact operation and maintenance of nuclear power plants in a number of ways; for example, (a) their neutronic properties can adversely affect the nuclear performance of the reactor: (b) their thermal resistance can cause elevated surface temperature on the fuel rods that may lead to material failure in the rod; (c) their radioactive decay results in work radiation exposure when they are redistributed throughout the reactor coolant system, in particular during power transients; (d) they complicate thorough inspection of irradiated nuclear fuel assemblies by both visual and eddy current methods; (e) deposits released from fuel rods tend to reduce visibility in the spent fuel pool, significantly delaying other work in the fuel pool during refueling outages; (f) once reloaded into the reactor on assemblies that will be irradiated a second or third time, they form an inventory of material that can be redistributed onto new fuel assemblies in a detrimental manner.
Axial offset anomaly (AOA) has been reported in pressurized water reactors (PWRs). AOA is a phenomenon in which deposits form on the fuel rod cladding due to the combination of local thermal-hydraulic conditions and primary-side fluid impurities characteristic of the reactor and the primary system. These deposits cause an abnormal power distribution along the axis of the core, reducing available margin under certain operating conditions. AOA has forced some power plants to reduce the reactor power level for extended periods.
Primary-side crud deposits are compositionally complex, containing four common constituents; nickel ferrite, nickel, nickel oxide, and zirconium oxide. Secondary circuit deposits consist primarily of magnetite (Fe3O4), with lesser concentrations of copper, zinc (as an oxide spinel or as the oxide), nickel (as the oxide or as nickel ferrite), and a host of minor mineral species that typically represent less than 2-3% of the deposit (by weight). These mineral species contain, among other elements, aluminum, silicon, calcium, magnesium, and manganese. Iron oxide is the predominant metal oxide contained in the metal-oxide/sludge formed in the secondary circuit nuclear steam generator.
The consequences resulting from the buildup of metal oxides within the secondary side of a steam generator are reduced steam output, thereby resulting in lost electrical output from the generating plant, increased water level fluctuations within the steam generator thereby resulting in lower steam and electrical output, and the initiation of corrosion deposits within the heat exchanger through the concentration of the dissolved chemical species from the secondary water. The corrosion within the secondary side of a pressurized nuclear steam generator ultimately may result in tube plugging and sleeving and the eventual loss of electrical output because of lost heat transfer or flow imbalances unless the steam generators themselves are replaced.
The deposits which form on both core and ex-core surfaces in the primary systems of nuclear plants, as well as on the secondary side of steam generators, are largely composed of crystalline solids. A crystalline material is one form of solid which exhibits a regularly ordered array of atoms in a lattice structure. Other solids which may exist in crud and deposits are amorphous (potentially some silicates or glass like species), and possibly some hydroxides or gel-like species. However, the vast majority of deposits are crystalline.
The deposits that adhere to surfaces on the primary and secondary side are thought to form by a number of mechanisms, including: (1) crystallization of soluble species from the coolant, (2) attachment of particulates that have been formed within the reactor coolant or secondary plant systems, or been introduced from outside the plant as impurities, (3) local transformation of existing deposits, and (4) oxidation or corrosion of the parent, underlying surface.
The process of crystallization involves two fundamental steps: (1) initial nucleation of a solute at a surface or within the solution, followed by (2) ongoing crystal growth by adsorption and incorporation of solute molecules at the crystal surfaces. The presence of a solute in a solution at concentrations above equilibrium (“super saturation”) is a major driving force for nucleation initiated crystallization, but crystallization can also occur from solutions that are not saturated if the formation of a solid phase, such as at a surface, is thermodynamically favorable. The external shape of a crystal is known as the crystal habit. Usually, crystal growth leads to the formation of crystal aggregates rather than single crystals, and the habit represents the appearance of the aggregate.
Crystal habit modifiers (CHM) are chemical additives that change the habit, or the shape, of crystals and in turn affect the behavior and properties of the crystals and crystal aggregates. CHMs are commonly used in the chemical industry to produce products with desirable crystalline structure, morphology, density, particle size, or surface area.
Many conventional crystal habit modifiers used in the chemical and pharmaceutical process industries may not be readily applied to a PWR plant environment, as they are incompatible with nuclear plant operations and chemistry specification limits.
Currently, the control of corrosion product deposition involves the minimization of the transport of corrosion products to fuel and steam generator (SG) surfaces and the mitigation of the deposition of corrosion products on fuel and SG surfaces. For example, dispersants are currently added to PWR secondary side water chemistry to mitigate the deposition of corrosion products on SG surfaces. No chemical additive or other technologies exist for positively modifying the crystalline structure of the fuel and SG deposits.